This invention relates broadly to improvements in nuclear fuel elements for use in the core of nuclear fission reactors and specifically, to improved nuclear fuel elements for use in boiling water reactors having improved stress corrosion cracking resistance and improved inner surface corrosion resistance.
Standard parts of nuclear reactors are the fuel elements forming the core of the reactor that contains the nuclear fuel. Although the fuel elements may assume any one of a number of geometric cross-sections, the elements are comprised of nuclear fuel enclosed by cladding. The cladding is ideally corrosion resistant, non-reactive and heat conductive. Coolant, typically demineralized water, flows in the flow channels that are formed between the fuel elements to remove heat from the core. One of the purposes of the cladding is to separate the nuclear material of the fuel from the coolant. Another purpose of the cladding is to minimize or prevent the radioactive fission products from contacting the coolant and thereby being spread throughout the primary cooling system. However, over time different cladding designs have failed by a number of failure mechanisms.
In order to accomplish these and other purposes, various materials and combinations of materials have been used in the cladding. The most common cladding materials include zirconium and alloys of zirconium, stainless steel, aluminum and its alloys, niobium and other materials. Of these, zirconium and its alloys have proven to be excellent materials for such purposes in water reactors because of material properties suited for cladding, including good heat conductivity, good strength and ductility, low neutron absorptivity and good resistance to corrosion.
One composite system utilizes an inner lining of stainless steel metallurgically bonded to zirconium alloy. The disadvantage of this system is that the stainless steel develops brittle phases that ultimately crack, allowing the by-products of the fission to contact the zirconium alloy cladding, initiating the deterioration of the zirconium alloy outer cladding. Furthermore, the stainless steel layer has a neutron absorption penalty of ten to fifteen times the penalty for a zirconium alloy of the same thickness. A solution to the problem of cladding failure is set forth in U.S. Pat. No. 3,969,186 which sets forth a composite consisting of refractory metals such as molybdenum, tungsten, rhenium, niobium and alloys of these materials in the form of a tube or foil of single or multiple layers or a coating on the internal surface of the cladding.
Still another solution to the problem is set forth in U.S. Pat. No. 4,045,288 that teaches the use of a composite cladding of zirconium alloy substrate with a sponge zirconium liner. The concept is that the commercially pure, soft, ductile zirconium liner minimizes the localized strain that the outer cladding is subject to. However, if a breach in the outer cladding should occur, allowing water and/or steam to enter the fuel rod, the zirconium liner tends to oxidize rapidly.
Yet another approach to the problem of cladding failure set forth in U.S. application Ser. No. 06/374,052 filed May 3, 1982, assigned to the assignee of the present application, and incorporated herein by reference, teaches using a composite cladding consisting of a dilute zirconium alloy inner liner metallurgically bonded to conventional cladding materials such as zirconium alloy claddings. The dilute zirconium alloy inner liner includes at least one metal alloyed with the zirconium selected from the group consisting of iron, chromium, iron plus chromium and copper. The amount of iron alloyed with the zirconium is from about 0.2% to about 0.3% by weight; the amount of chromium is from about 0.05% to about 0.3% by weight; the total amount of iron plus chromium is from about 0.15% to about 0.3% by weight and wherein the ratio of the weights of iron to chromium is in the range of from about 1:1 to about 4:1; and wherein the amount of copper is from about 0.02% to about 0.2% by weight.
While advances have been made in the area of improving the performance of claddings, corrosion and brittle splitting of the cladding due to interactions of the cladding, the nuclear fuel, the fission products and the coolant continues to be a problem even with the improved systems.
A particularly effective nuclear fuel element is comprised of a central core of a nuclear fuel material. The nuclear material may be any radioactive materials, such as the well known radioactive materials of uranium, plutonium, thorium and mixtures thereof.
The central core of nuclear fuel material is surrounded by an elongated composite cladding comprised on an inner metallic barrier and an outer metallic tubular portion. The outer portion of the cladding is unchanged in design and function from the previous practices utilized in the nuclear reactor arts. The outer metallic tubular portion remains the standard, well-known materials conventionally used in cladding, and in particular, as outer portions of composite claddings. The outer metallic portion is selected from the group consisting of zirconium and its alloys, stainless steel, aluminum and its alloys, niobium and magnesium alloys.
The inner metallic barrier is zirconium in which the amount of Fe is microalloyed with the zirconium in a controlled amount of from about 850-2500 parts per million by weight (ppm). The inner metallic barrier is metallurgically bonded to the outer metallic tubular portion, but unlike the outer metallic portion when comprised of zirconium or its alloys, is alloyed only with carefully controlled amounts of iron. Trace elements in an amount so as not to affect the character and nature of the inner metallic barrier may be present.
Surprisingly, by carefully controlling the amount of iron present in the zirconium, it has been discovered that the inner metallic barrier not only has greatly improved corrosion resistance over previous claddings and barriers but also improved stress corrosion cracking resistance, while the other important characteristics of the zirconium inner metallic barrier are unaffected. The barrier is ductile, compatible with the outer metallic tubular portion, but has low neutron absorptivity, yet is highly resistant to radiation hardening while maintaining good heat transfer characteristics.
It is believed that the present invention improves the ability of the fuel element to operate normally in the failed condition, that is, with the failure of the outer cladding due to primary defects developed as a result of stress corrosion or fretting, but without developing secondary long axial cracks along the inner barrier. The inner metallic barrier has sufficient corrosion resistance such that it will continue to provide an effective barrier when exposed to the nuclear fuel and the by-products of nuclear fission as well as the coolant, which may include demineralized water, steam and/or moderators. The life expectancy of the fuel element is increased, even after failure of the outer cladding, due to the ability of the inner metallic barrier to slow down the formation of corrosion products (hydrides) upon contact with coolant.
Other features and advantages of the present invention will be apparent from the following description and the preferred embodiment, taken in conjunction with the accompanying drawings which illustrate, by way of example, the principles of the invention.